ML16263A319

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MRP-227-A Applicant/Licensee Action Item 7 Analysis, Topical Report, ANP-3479NP, Revision 0.
ML16263A319
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Issue date: 08/31/2016
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Attachment 2 "MRP-227-A Applicant/Licensee Action Item 7 Analysis for Three Mile Island Unit 1, Topical Report," ANP-3479NP, Revision 0, Non-Proprietary Version

Controlled Document 0414-12-F04 (Rev. 001, 03/10/2016)

A AREVA ANP-3479NP MRP-227-A Applicant/Licensee Action Revision 0 Item 7 Analysis for Three Mile Island Unit 1 Topical Report August 2016 AREVA Inc.

(c) 2016 AREVA Inc.

Controlled Document ANP-3479NP Revision 0

  • Copyright © 2016 AREVA Inc.

All Rights Reserved

Controlled Document AREVA Inc. ANP-3479NP Revision 0 MRP-227-A Applicant/Licensee Action Item 7 Analysis for Three Mile Island Unit 1 Topical Report Page i Nature of Changes Section(s)

Item or Page(s) Description and Justification 1 All Initial Issue

ntrolled Document AREVA Inc. ANP-3479NP Revision 0 MRP-227-A ApplicanULicensee Action Item 7 Analysis for Three Mile Island Unit 1 Topical Report Page ii Contents Page

1.0 INTRODUCTION

AND PURPOSE .................................................................... 1-1 2.0 METHODOLOGY .............................................................................................. 2-1 2.1 WCAP-17096 Methodology Applicability ................................................. 2-1 2.2 MRP-227-A Suggested Methodologies ................................................... 2-1 2.3 Methodology Utilized for TMl-1 ............................................................. :.2-2 3.0 CRGT SPACER CASTINGS ............................................................................. 3-1 3.1 Background ............................................................................................. 3-1 3.1.1 Description of the Component Item .............................................. 3-1 3.2 Evaluation Inputs .................................................................................... 3-3 3.2.1 Flaw Size ..................................................................................... 3-3 3.2.2 Degraded Material Properties ...................................................... 3-4 3.2.3 Distortion Evaluation .................................................................... 3-5 3.3 Evaluation ............................................................................................... 3-6 3.3.1 Likelihood of Failure ................................................................ ~ .... 3-6 3.3.2 Effect of Failure on Functionality .................................................. 3-7 3.4 Conclusions ............................................................................................ 3-9 4.0 IMI GUIDE TUBE SPIDER CASTINGS ............................................................. 4-1 4.1 Background ............................................................................................. 4-1 4 .1 .1 Description of the Component Item .............................................. 4-1 4.2 Evaluation Inputs .......................................................................... :......... 4-3 4.2.1 Likelihood of Fabrication and Service-Induced Flaws .................. 4-4 4.2.2 Driving Force for Crack Extension ................................................ 4-4 4.2.3 Irradiated Fracture Toughness ..................................................... 4-5 4.3 Evaluation ............................................................................................... 4-7 4.3.1 Likelihood of Failure ..................................................................... 4-7 4.3.2 Impact of Fractured Spider Casting on Functionality .................... 4-8 4.4 Conclusions ............................................................................................ 4-9 5.0 VENT VALVE RETAINING RINGS ....................................................... ~ ............ 5-1 5.1 Background .......................................... ,.................................................. 5-1 5.1.1 Description of the Component Item .............................................. 5-1 5.2 Evaluation Inputs .................................................................................... 5-3 5.2.1 Flaw Size ..................................................................................... 5-3 5.2.2 Degraded Material Properties ...................................................... 5-3

Controlled Document AREVA Inc. ANP-3479NP Revision 0 MRP-227-A Applicant/Licensee Action Item 7 Analysis for Three Mile Island Unit 1 Topical Report Page iii 5.2.3 Stresses ....................................................................................... 5-4 5.3 Evaluation ...................*............................................................................ 5-4 5.3.1 Likelihood of Failure ..................................................................... 5-5 5.3.2 Effect of Failure on Functionality .................................................. 5-5 5.4 Conclusions ............................................................................................ 5-6 6.0 OVERALL CONCLUSIONS .. ,............................................................................ 6-1

7.0 REFERENCES

.................................................................................................. 7-1

Controlled Document AREVA Inc. ANP-3479NP Revision 0 MRP-227-A Applicant/Licensee Action Item 7 Analysis for Three Mile Island Unit 1 Topical Report Page iv Figures Figure 3-1: Schematic Representation of Typical Control Rod Guide Tube with Spacer Castings for B&W-Designed PWRs (One of 69 CRGTs Shown) (1) ........................................ :........................................................ 3-2 Figure 4-1: A Schematic Representation of Typical IMI Guide Tube Spider Castings in B&W-Designed PWRs (1) ....................................................... 4-2 Figure 5-1: A Schematic Representation of the Vent Valve Retaining Rings in B&W-Designed PWRs (1) (View from Inside the Core Support Shield) ..... 5-2

Controlled Document AREVA Inc. ANP-3479NP Revision 0 MRP-227-A Applicant/Licensee Action Item 7 Analysis for Three Mile Island Unit 1 Topical Report Pagev Nomenclature Acronym Definition A/LAI ApplicanULicensee Action Item ASME American Society of Mechanical Engineers ASTM American Society of Testing and Materials B&W Babcock & Wilcox BWRVIP Boiling Water Reactor Vessel and Internals Program CASS Cast Austenitic Stainless Steel CMTR Certified Material Test Report CRA Control Rod Assembly CRGT Control Rod Guide Tube css Core Support Shield EFPY Effective Full Power Year EPRI Electric Power Research Institute FIV Flow-Induced Vibration l&E Inspection and Evaluation IE Irradiation Embrittlement IMI lncore Monitoring Instrumentation LOCA Loss of Coolant Accident LR License Renewal LWR Light Water Reactor MRP Materials Reliability Program NOE Non-Destructive Evaluation NRG Nuclear Regulatory Commission NSSS Nuclear Steam Supply System PH Precipitation-Hardenable PT Dye Penetrant Testing PWR Pressurized Water Reactor RT Radiographic Testing RV Reactor Vessel sec Stress Corrosion Cracking SER Safety Evaluation Report SSE Safe Shutdown Earthquake TE Thermal Aging Embrittlement TMl-1 Three Mile Island Unit 1 U.S. United States.

UT Ultrasonic Testing VT Visual Testing

Controlled Document AREVA Inc.

ANP-3479NP Revision 0 MRP-227-A ApplicanULicensee Action Item 7 Analysis for Three Mile Island Unit 1 Topical Report Page vi ABSTRACT The purpose of this document is to summarize the analyses performed for the applicable component items at Three Mile Island Unit 1 to complete applicant/licensee action item #7 from MRP-227-A for Three Mile Island Unit 1. The summary includes a discussion of the purpose; the methodology utilized; a summary of the background, evaluation inputs, evaluation, and conclusion for each component item; and an overall conclusion.

Controlled Document AREVA Inc. ANP-3479NP Revision 0 MRP-227-A Applicant/Licensee Action Item 7 Analysis for Three Mile Island Unit 1 Topical Report Page 1-1

1.0 INTRODUCTION

AND PURPOSE The Electric Power Research Institute (EPRI) Materials Reliability Program (MRP) developed inspection and evaluation (l&E) guidelines in document MRP-227-A (1) for managing long-term aging of pressurized water reactor (PWR) reactor vessel (RV) internal components. Specifically, the l&E guidelines are applicable to RV internal structural components; they do not address fuel assemblies, reactivity control assemblies, or welded attachments to the RV. The l&E guidelines concentrate on eight aging degradation mechanisms and their aging effects, such as loss of fracture toughness. The l&E guidelines define requirements for inspections that will allow owners of PWRs to demonstrate that the effects of aging degradation are adequately managed for the period of extended operation. These guidelines contain mandatory and needed requirements and an implementation schedule for nuclear units employing Babcock and Wilcox (B&W) nuclear steam supply systems (NSSSs) currently operating in the United States (U. S.).

MRP-227-A includes a safety evaluation report (SER) prepared by the U.S. Nuclear Regulatory Commission (NRC). The U.S. NRC staff determined whether the guidance contained in the report provided reasonable assurance that the l&E guidelines ensured that the RV internal components will maintain their intended functions during the period of extended operation. From the determination, seven topical report conditions and eight plant-specific applicant/licensee action items (A/LAls) were contained in the SER to alleviate issues and concerns of the NRC staff. The plant-specific A/LAls address topics related to the implementation of MRP-227 that could not be effectively addressed on a generic basis in MRP-227. The seventh A/LAI (A/LAI 7) addresses NRC staff concerns regarding thermal aging embrittlement (TE) and irradiation embrittlement (IE).

During the performance of this A/LAI, three component items were identified as requiring further aging management for Three Mile Island Unit 1 (TMl-1) based on material type. A fourth component item, the vent valve bodies, were also identified as being fabricated from cast austenitic stainless steel (CASS); however they were not

Controlled Document AREVA Inc. ANP-3479NP Revision 0 MRP-227-A ApplicanULicensee Action Item 7 Analysis for Three Mile Island Unit 1 Topical Report Page 1-2 evaluated since the vent valve bodies in the currently installed [

] at TMl-1 have a ferrite content of 20% or less. Based on a records search performed in 2010, [

] One (1) vent valve assembly was replaced as of the Fall 2015 outage. Data from the CMTR of the vent valve body that was installed in 2015 was used to calculate the ferrite content and was determined to be less than 20%. Therefore the vent valve bodies are screened out as being susceptible to TE or IE.

The three component items applicable to A/LAI #7 for TMl-1 are listed below:

  • Control Rod Guide Tube (CRGT) Spacer Castings (Grade CF-3M) o Screened as potentially susceptible to TE, but not IE
  • lncore Monitoring Instrumentation (IMI) Guide Tube Spiders (Grade CF-8) o Screened as potentially susceptible to IE, but not TE
  • Vent Valve Retaining Rings (Type 15-5 precipitation-hardenable [PH])

o Screened as potentially susceptible to TE, but not IE The purpose of this document is to summarize analyses performed for these three component items for Exelon's TMl-1. This document will fulfill the A/LAI for these component items; that is, to report the results of the plant-specific analysis for TMl-1 developed to demonstrate that the component items will maintain their functionality dudng the period of extended operation, considering the loss of fracture toughness due to TE and/or IE (whichever is applicable).

The methodology used to evaluate all three components items is similar and is illustrated in Section 2.0 . Each component item has its own section (CRGT Spacer

1--~

Controlled Document AREVA Inc. ANP-3479NP Revision 0 MRP-227-A ApplicanULicensee Action Item 7 Analysis for Three Mile Island Unit 1 Topical Report Page 1-3 Castings - Section 3.0, IMI Guide Tube Spider Castings - Section 4.0, Vent Valve Retaining Rings - Section 5.0).

Information considered by AREVA to be proprietary is marked with brackets: [ ]

Controlled Document AREVA Inc. ANP-3479NP Revision 0 MRP-227-A ApplicanULicensee Action Item 7 Analysis for Three Mile Island Unit 1 Topical Report Page 2-1 2.0 METHODOLOGY The purpose of this section is to provide various potential methodologies and identify the ultimate methodology used to evaluate the component items for TMl-1.

2.1 WCAP-17096 Methodology Applicability WCAP-17096 (2), including proposed edits (3), provides a methodology for developing evaluation procedures to assess the functional impacts of degradation in component items with "observed relevant conditions." As will be discussed below, the basis for why these component items are not expected to fail [ ]

is generally part of the methodology used herein to justify that these component items will be expected to maintain their functionality through the period of extended operation.

Therefore, the WCAP-17096 methodology ( )

2.2 MRP-227-A Suggested Methodologies As described in A/LAI 7, to address the NRC staff concerns regarding TE and IE of potentially susceptible materials, applicants/licensees are required to perform a plant-specific analysis or evaluation demonstrating that certain component items will maintain their functionality during the period of extended operation. Per MRP-227-A, possible acceptable approaches may include, but are not limited to:

  • Functionality analyses for the set of like components or assembly-level functionality analyses, or
  • Component level flaw tolerance evaluation justifying that the MRP-227 recommended inspection technique(s) can detect a structurally significant flaw for the component in question, taking into account the reduction in fracture toughness due to IE and TE; or

Controlled Document AREVA Inc. ANP-3479NP Revision 0 MRP-227-A Applicant/Licensee Action Item 7 Analysis for Three Mile l.sland Unit 1 Topical Report Page 2-2

  • For CASS, if the application of applicable screening criteria for the component's material demonstrates that the components are not susceptible to either TE or IE, or the synergistic effects of TE and IE, then no other evaluation would be necessary. For assessment of CASS materials, the licensees or applicants for license renewal (LR) may apply the criteria in the NRG letter of May 19, 2000, "License Renewal Issue No. 98-0030, Thermal Aging Embrittlement of Cast Stainless Steel Components" (4) as the basis for determining whether the CASS materials are susceptible to the TE mechanism.

2.3 Methodology Utilized for TMl-1 As WCAP-17096 (

] the methodology discussed below will be applied:

  • Identify appropriate inputs for the evaluation, [

]

  • Utilize available information to determine if failure is likely or unlikely to occur
  • Determine effect of failure on functionality
  • Conclude whether componentitems are expected to maintain their functionality through the period of extended operation

Controlled Document AREVA Inc. ANP-3479NP Revision 0 MRP-227-A ApplicanULicensee Action Item 7 Analysis for Three Mile Island Unit 1 Topical Report Page 3-1 3.0 CRGT SPACER CASTINGS This section summarizes the analysis performed of the TMl-1 CRGT spacer castings to fulfill A/LAI #7 from MRP-227-A.

3.1 Background MRP-227-A provides l&E guidelines for the various component items including the CRGT spacer castings, which are considered a "Primary" component item in MRP-227-A. The l&E guidelines specify applicability, effect and mechanism, expansion link, examination method/frequency, and examination coverage.

3.1.1 Description of the Component Item This section contains an abbreviated description, including a short description of the functionality, consequence of failure, and operating experience of the component items.

  • The plenum assembly (upper internals) contains 69 vertical CRGT assemblies. Inside of each guide housing is a brazement subassembly consisting of ten parallel spacer castings brazed to twelve perforated vertical rod guide tubes and 4 pairs of vertical rod tube guide sectors (see Figure 3-1). There are a total of 690 spacer castings in the TMl-1 RV internals. The CRGT spacer castings are made from ASTM A 351-65, Grade CF-3M castings.

Controlled Document AREVA Inc. ANP-3479NP Revision 0 MRP-227-A Applicant/Licensee Action Item 7 A nalysis for Three Mile Island Unit 1 Topical Report Page 3-2 Figure 3-1: Schematic Representation of Typical Control Rod Guide Tube with Spacer Castings for B&W-Designed PWRs (One of 69 CRGTs Shown) (1)

During normal operation , most of the control rod assembly (CRA) is positioned with in the CRGT. In the event of a reactor trip or a rod movement command from the control room , the CRAs pass through the path provided by the brazement into , or out of, the fuel assemblies . The outer pipe portion of the CRGTs is the structural support for the rod guide brazement, and also provides a structural connection between the upper grid assembly and the plenum cover in the upper internals.

There are open ings in the lower region of the pipe to allow some of the fluid entering the CRGT assembly from the core to exit to the plenum region . The rema inder of the CRGT assembly has [ ]

Controlled Document AREVA Inc. ANP-3479NP Revision 0 MRP-227-A ApplicanULicensee Action Item 7 Analysis for Three Mile Island Unit 1 Topical Report Page 3-3 The function of the CRGT spacer castings is to provide structural support and alignment to the [ ] vertical rod guide tubes and [ ] vertical rod guide sectors within each CRGT assembly. The CRA consists of a control rod spider and control rods that travel vertically within the rod guide brazement. [

] The CRA is guided by the brazement subassembly over the entire range of the vertical withdrawal path . In addition , the rod guide tubes limit reactor coolant cross-flow on the control rods, which limits flow-induced vibration (FIV) . The spacer castings do not have a core support function ; however, they do have a safety function relative to control rod alignment, insertion , and reactivity issues. Degradation of the spacer castings could result in degradation of the unit's shutdown capability by hindering the insertion of the control rods into the core within the normal anticipated time .

Appendix A of MRP-227-A indicates that no failures of CASS materials due to TE in the PWR RV internals have been reported. Additionally , no known failures of CASS materials due to embrittlement have been reported by the industry.

3.2 Evaluation Inputs This section will describe the quantitative inputs for the evaluation , such as flaw size ,

degraded material properties, and stresses.

3.2.1 Flaw Size As indicated by MRP-227-A, the CRGT spacer castings are not screened as potentially susceptible to service induced flaws (i.e., irradiation-assisted stress corrosion cracking

[IASCC] , SCC , or fatigue). Therefore , the following section will focus on the potential for flaws in the 'as-built' condition from the manufacturing process . The non-destructive evaluation (NOE) methods used to examine the component items prior to their in-service time at TMl-1 are summarized below.

Controlled Docume AREVA Inc. ANP-3479NP Revision 0 MRP-227-A Applicant/Licensee Action Item 7 Analysis for Three Mile Island Unit 1 Topical Report Page 3-4 Review of the available CMTRs for the CRGT spacer castings indicated [

]

However, it is reasonable to assume that the CRGT spacer castings used in-service at TMl-1 ( ] and therefore [

]

3.2.2 Degraded Material Properties

[ ] of the CRGT spacer castings at TMl-1 exceed the screening criteria for TE and are therefore considered susceptible. The time to reach saturation in the reduction of impact properties was investigated for the susceptible CRGT spacer castings. It was determined that a saturation value of impact energy, and, correspondingly, fracture toughness, [

] Therefore, for the susceptible CRGT spacer castings (with high ferrite contents exceeding the screening criteria), [

]

l Controlled Document AREVA Inc. ANP-3479NP Revision 0 MRP-227-A ApplicantiLicensee Action Item 7 Analysis for Three Mile Island Unit 1 Topical Report Page 3-5 3.2.3 Distortion Evaluation An evaluation was performed to determine the amount of distortion allowed that will still permit the control rod spider to freely pass through the brazement sub-assembly. The acceptance criterion was then used in an analysis of the brazement subassembly to evaluate the conditions required to lead to a restricted guide path. The brazement subassembly analysis only evaluated the [

]

The conclusion of the analysis was [

]

Controlled Document AREVA Inc. ANP-3479NP Revision 0 MRP-227-A Applicant/Licensee Action Item 7 Analysis for Three Mile Island Unit 1 Topical Report Page 3-6 3.3 Evaluation The results of the methodology utilized are organized into several conclusions as discussed in the following sections.

3.3.1 Likelihood of Failure Between 2012 and 2014, three B&W units performed visual testing (VT)-3 examinations of the CRGT spacer castings, per the guidance in MRP-227-A. These visual examinations, with 100% coverage of accessible surfaces at each of the four CRGT spacer casting screw locations, revealed no recordable indications. Furthermore, no known failures of CASS materials due to embrittlement have been reported by the industry. This is confirmation that [

]

Due to the [ ] being present in the material [

]

Considering the above discussion, and the fact that the [

] failure of the CRGT spacer casting material during the period of extended operation is unlikely.

Controlled Document AREVA Inc.

ANP-3479NP Revision 0 MRP-227-A ApplicanULicensee Action Item 7 Analysis for Three Mile Island Unit 1 Topical Report Page 3-7 3.3.2 Effect of Failure on Functionality

[

] However, the postulated occurrence is not considered a credible scenario as described below, The reported stress distribution reinforces the premise that [

] is not expected. The stress analysis of the spacer castings verifies that the highest stresses in the castings are located at the four screw locations and are reverse bending stresses around the circumference.

[

] This analysis also leads to the conclusion that .[

] In particular, stresses in the [

]

Therefore, the first failure reduces the probability of the second failure. This data directly reinforces the premise that if a failure were to occur, [

1

Controlled Document AREVA Inc. ANP-3479NP Revision 0 MRP-227-A ApplicanULicensee Action Item 7 Analysis for Three Mile Island Unit 1 Topical Report Page 3-8 Drop-time testing of the CRAs is performed at the beginning of each cycle per TMl-1 technical specifications. Historically, the rod drop-times are somewhat uniform and are easily within the required limit. When an unusual drop-time is encountered, the utility normally investigates the possible cause. To date, slow trip times have always been attributed to unusual fuel bow or issues with the control rod drive mechanism.

Additionally, the safety analysis already considers that the maximum worth rod may not trip on demand.

An analysis of the brazement sub-assembly, including postulated failed castings, demonstrates that the deformation of a casting with a single failure at any screw location would be acceptable and not restrict the control rod guide path. Multiple single failures of castings in the same brazement sub-assembly have also been shown to be acceptable. The redundant features of the [

] The rod guide tubes and rod guide sectors have [ l In addition, [

] due to the CRGT spacer casting geometry and [

] Spacer casting failures at multiple locations would have to occur accompanied by a complete braze failure in the vicinity of the casting failures to generate loose parts. Therefore it is concluded that loose parts are not a credible risk.

Controlled Document AREVA Inc. ANP-3479NP Revision 0 MRP-227-A Applicant/Licensee Action Item 7 Analysis for Three Mile Island Unit 1 Topical Report Page 3-9 3.4 Conclusions Cast austenitic stainless steel materials are known to be potentially susceptible due to TE after exposure at PWR RV internals temperatures for long periods of time, especially those containing higher levels of ferrite and molybdenum. Studies show that saturation with respect to the reduction of room temperature impact energy, and correspondingly fracture toughness, [

] The good operating experience demonstrates that normal operating stresses are expected to be minimal. The CRGT spacer castings not exceeding the screening criteria are not considered susceptible to TE.

Additionally, [ ] for the CRGT spacer castings [

] The stress analysis of the spacer castings [

]

Analyses show [

] the monitoring of control rod drop times is a normal surveillance requirement, and the utility normally investigates abnormal control rod drop times. [ ]

Controlled Document AREVA Inc. ANP-3479NP Revision 0 MRP-227-A ApplicanULicensee Action Item 7 Analysis for Three Mile Island Unit 1 Topical Report Page 3-10 The function of the brazement sub-assembly, of which the CRGT spacer castings are a part, is currently monitored by periodically verifying the control rod movement and control rod drop-time. In addition, the aging management plan has implemented a new visual examination to verify that critical regions of the spacer casting have not failed due to random original defects. The combination of the original licensing basis surveillance, that has not been altered, and the additional visual verification of intact spacer castings provide assurance that the spacer castings will continue to perform their function for the period of extended operation.

Based on the discussion above, it is concluded that the CRGT spacer castings are expected to maintain functionality during the period of extended operation.

Controlled Document AREVA Inc. ANP-3479NP Revision 0 MRP-227-A ApplicanULicensee Action Item 7 Analysis for Three Mile Island Unit 1 Topical Report Page 4-1 4.0 IMI GUIDE TUBE SPIDER CASTINGS This section summarizes the analysis performed of the TMl-1 IMI guide tube spider castings to fulfill A/LAI 7 from MRP-227-A.

4.1 Background MRP-227-A provides l&E guidelines for the various component items including the IMI guide tube spider castings, which are considered a "Primary" component item in MRP-227-A. The l&E guidelines specify applicability, effect and mechanism, expansion link, examination method/frequency, and examination coverage.

4.1.1 Description of the Component Item This section contains an abbreviated description, including a short description of the functionality, consequence of failure, and operating experience of the component items.

The IMI guide tube spider castings are part of the lower internals assembly. Fifty-two IMI guide tube assemblies provide support and protection for the IMI along the path from the RV bottom head IMI nozzles, through the lower internals, and into the instrument tubes in the fuel assemblies (see Figure 4-1).

Controlled Document AREVA Inc. ANP-3479NP Revis ion 0 MRP-227-A ApplicanULicensee Action Item 7 Analysis for Three Mile Island Unit 1 Top ica l Report Page 4-2 IMI Guide Tube Spider Castings Figure 4-1: A Schematic Representation of Typical IMI Guide Tube Spider Castings in B&W-Designed PWRs (1)

The IMI guide tube spider castings are ASTM A351-65 Grade CF-8 material and resemble a four eared butterfly nut. The IMI guide tube spider casting has a center hub with four integral "L" shaped legs extending outward . The inner diameter of each IMI guide tube spider casting center hub is chrome plated . Each of the 52 IMI guide tube spider castings is custom machined to fit within the lower grid rib section .

Each of the four IMI guide tube spider casting legs is fillet welded to the walls of the lower grid rib section . The welds are a stainless steel filler metal (ER 308/308L) . The tip at the upper end of the IMI guide tube slides inside the chrome-plated center hub of the IMI guide tube spider casting . The relatively [

] to accommodate the axial expansion of the IMI guide tube.

Controlled Document AREVA Inc. ANP-3479NP Revision 0 MRP-227-A Applicant/Licensee Action Item 7 Analysis for Three Mile Island Unit 1 Topical Report Page 4-3 The lower end of the IMI guide tube is solidly welded to the flow distributor head (in some locations with the use of gussets) with additional support provided at its midsection via a threaded guide tube nut to the IMI guide support plate .

The function of the IMI guide tube spider casting is to provide lateral restraint for the IMI guide tube and the function of the spider fillet welds is to hold the IMI guide tube spider casting in place . The IMI guide tube provides a continuous protected guide path for the in-core monitoring instrumentation from their entry into the RV, through the RV instrumentation nozzles, to the entrance into the fuel assembly instrument guide tube .

[

] Loss of function of the in-core monitoring instrument guide path would be a sufficient misalignment at the fuel assembly instrument tube entrance to prohibit entry of the in-core instrument. In addition, failure of the guide path could result in wear of the IMI sheath due to FIV and therefore would be considered a loss of function of the IMI.

Appendix A of MRP-227-A indicates no cracking has been reported in PWR RV internals attributed to the embrittlement of CASS materials. Cast stainless steels are used extensively in pressure-boundary components such as piping components , valve bodies, and pump casings. However, no cases of embrittlement requiring corrective action have been reported by the industry as of 2015 .

4.2 Evaluation Inputs This section will describe the quantitative inputs for the evaluation , such as flaw size ,

degraded material properties , and stresses.

Controlled Document AREVA Inc. ANP-3479NP Revision 0 MRP-227-A Applicant/Licensee Action Item 7 Analysis for Three Mile Island Unit 1 Topical Report Page 4-4 4.2.1 Likelihood of Fabrication and Service-Induced Flaws

[

] Information regarding the [

] is limited to the severity of subsurface defects deemed acceptable by the American Society of Mechanical Engineers (ASME) Code for castings in pressure boundary applications.

Additionally, service-induced flaws such as those resulting from SCC, IASCC, and fatigue, were evaluated and screened out for the TMl-1 IMI guide tube spider castings.

4.2.2 Driving Force for Crack Extension A structural analysis evaluated the design configuration for the TMl-1 IMI guide tube spider castings and considered two intact loading configurations:

Configuration 1: steady state reactor coolant flow with accelerations from safe shutdown earthquake (SSE) and loss of coolant accident (LOCA) events Configuration 2: [

1 This structural analysis shows that the stress in the TMl-1 IMI guide tube spider castings is [ ] In summary, loads capable of driving a [

] Loads from [

1

Controlled Document

. AREVA Inc. ANP-3479NP

  • Revision 0 MRP-227-A ApplicanULicensee Action Item 7 Analysis for Three Mile Island Unit 1 Topical Report Page 4-5 4.2.3 Irradiated Fracture Toughness The IMI guide tube spider castings retain sufficient fracture toughness to be best characterized by elastic-plastic measures of fracture toughness. Two elastic-plastic measures of fracture toughness are used in this report: J1c and J at a crack extension of 2.5 mm (hereafter "J2.smm"). The parameter J characterizes the crack driving force based on an integration of loading work per unit volume (e.g., strain energy density for elastic bodies) around a crack front. J 1c characterizes the crack driving force just prior to the onset of significant stable tearing crack extension. J2.smm characterizes the crack driving force required to achieve a crack extension of 2.5mm.

4.2.3.1 Fracture Toughness Characterized by J1c There is a paucity of fracture toughness data available for CASS material, particularly in the [ ] relevant to the TMl-1 IMI guide tube spider castings. No measured fracture toughness properties were identified in this task for CF-8 materials irradiated in light water reactors (LWRs). Reference (5) reports fracture toughness properties (measured at room temperature) for a CF-8 material irradiated at 325°C in a fast-breeder reactor between roughly 6 and 12 dpa. The .Reference (5) data are summarized in Figure 54 of Reference (6) as "CF-8 (Burke et al.)", including an additional measurement at 19 dpa. A lower bound to this data, determined per engineering judgment, suggests that the [

]

based on the fracture toughness categorizations described in Reference (6). Figure 61 of Reference (6) shows that for irradiated materials [

]

Controlled Document AREVA Inc. ANP-3479NP Revision 0 MRP-227-A Applicant/Licensee Action Item 7 Analysis for Three Mile Island Unit 1 Topical Report Page 4-6 The key point from this discussion is that the TMl-1 IMI guide tube spider casting [

]

4.2.3.2 Fracture Toughness Characterized by J2.Smm The NRC has adopted J2.smm =255 kJ/m 2 as a conservative criterion for piping materials to differentiate between non-significant and potentially significant reduction in fracture toughness for CASS subject to thermal aging embrittlement (4). A joint Boiling Water Reactor Vessel and Internals Program (BWRVIP)/MRP Working Group on CASS has compiled information of J2.smm for irradiated CF-8 materials as a function of neutron exposure. Most of this J2.smm data is from the same CF-8 testing from which J1c was 2

discussed in Section 4.2.3.1 and shows that J2.smm = 255 kJ/m is not reached until about 3.3 dpa. In addition, the BWRVIP/MRP Working Group reported example calculations of J2.smm for RV internals components with large flaws. These calculations show that is that a crack driving force of Japplied 255 kJ/m 2 is unlikely to be achieved in

=

RV internals components, adding further conservatism to the use of J2.smm = 255 kJ/m 2 .

[ ] (the lower bound estimate for J2.smm =

255 kJ/m 2 ) [

] There is also margin between Japplied =255 kJ/m 2 and the RV internals loading conditions.

The key point from this discussion [

] the NRC's conservative screening criterion of J2.smm =255 kJ/m 2 . Thus, the reduction in fracture toughness due to IE is not considered significant.

Controlled Document AREVA Inc. ANP-3479NP Revision 0 MRP-227-A ApplicanULicensee Action Item 7 Analysis for Three Mile Island Unit 1 Topical Report Page 4-7 4.3 Evaluation The results of the methodology utilized are organized into several conclusions as discussed in the following sections.

4.3.1 Likelihood of Failure Three parameters must be considered to evaluate the likelihood of failure due to reduced fracture toughness: 1) likelihood for flaw to be present, 2) driving force for crack extension and 3) material fracture toughness. Considering each of these parameters in turn, it is unlikely that a TMl-1 IMI guide tube spider casting will fail due to IE.

1) It is unlikely [

] (Section 4.2.1).

2) The dominant driving force for flaw extension in the TMl-1 IMI guide tube spider casting [ ] As a flaw grows, the driving force [

] (Section 4.2.2).

3) The TMl-1 IMI guide tube spider castings [

] relative to [ ] austenitic stainless steel and the NRC's conservative screening criterion of J2 .smm =255 kJ/m2 (Section 4.2.3).

The worst effect of reduced fracture toughness is expected [

] in the IMI guide tube spider casting. The flaw extension would be

[

] The unlikely, but more severe, case of [

] IMI guide tube spider casting leg is evaluated in Section 4.3.2.

Controlled Document AREVA Inc. ANP-3479NP Revision 0 MRP-227-A Applicant/Licensee Action Item 7 Analysis for Three Mile Island Unit 1 Topical Report Page 4-8 4.3.2 Impact of Fractured Spider Casting on Functionality The structural and FIV analyses previously described in Section 4.2.2 evaluated a series of [ ] of IMI guide tube spider castings [ ]

The structural analysis results show that stresses [

] and accelerations from SSE and LOCA [ ] When [

] the structural analysis results show that the maximum stress [ ] The FIV analysis results show that [

] Based on these analysis results, [ ] IMI guide tube spider casting [

] on that IMI guide tube spider casting.

The function of the IMI guide tube spider castings is to provide lateral restraint for the IMI guide tubes and the function of the fillet welds is to hold the IMI guide tube spider casting in place. Significant degradation of the IMI guide path (potentially due to contributions from degraded IMI guide tube spider castings) could result in misalignment at the fuel assembly instrument tube entrance, prohibiting free entry or withdrawal of the IMI itself, or could result in wear of the IMI.

[ ] is not expected to affect the function of the IMI guide tube spider casting for the following reasons:

1. The IMI guide tube spider casting provides [

]

Controlled Document AREVA Inc.

ANP-3479NP Revision 0 MRP-227-A ApplicanULicensee Action Item 7 Analysis for Three Mile Island Unit 1 Topical Report Page 4-9

2. The FIV analysis shows degraded configurations of the IMI guide tube spider casting

[ ] In addition, this analysis

[ ] of the IMI guide tube [

] of the IMI guide tube spider casting.

4.4 Conclusions The TMl-1 IMI guide tube spider castings are [

] from TMl-1 IMI guide tube spider casting fabrication records.

An TMl-1 specific neutron fluence evaluation [

] However, [ ] that the TMl-1 IMI guide tube spider castings [ ]

1. It is not likely that [

] in the IMI guide tube spider casting.

2. The dominant driving force for flaw extension in the TMl-1 IMI guide tube spider casting [

]

3. The TMl-1 IMI guide tube spider castings retain [

] relative to [

] the NRC's conservative screening criterion of J2.smm =255 2

kJ/m .

Controlled Document AREVA Inc. ANP-3479NP Revision 0 MRP-227-A Applicant/Licensee Action Item 7 Analysis for Three Mile Island Unit 1 Topical Report Page 4-10

[

1 Therefore, the IMI guide tube spider castings are expected to perform their function for the period of extended operation.

Controlled Document AREVA Inc. ANP-3479NP Revision 0 MRP-227-A ApplicanULicensee Action Item 7 Analysis for Three Mile Island Unit 1 Topical Report Page 5-1 5.0 VENT VALVE RETAINING RINGS This section summarizes the analysis performed of the TMl-1 vent valve retaining rings to fulfill A/LAI #7 from MRP-227-A.

5.1 Background MRP-227-A provides l&E guidelines for the various component items including the vent valve retaining rings, which are considered a "Primary" component item in MRP-227-A.

The l&E guidelines specify applicability, effect and mechanism, expansion link, .

examination method/frequency, and examination coverage.

5.1.1 Description of the Component Item This section contains an abbreviated description, including a short description of the functionality, consequence of failure, and operating experience of the component items.

TMl-1 has eight vent valves installed in the core support shield (CSS) cylinder. Each vent valve is mounted in a vent valve mounting ring (also called vent valve nozzle) which is welded into the CSS cylinder. For all normal operating conditions, the vent valve is closed. In the event of a pipe rupture in the RV inlet pipe, the valve opens to permit steam generated in the core to flow directly to the break. This permits the core to be flooded and adequately cooled when emergency core coolant is *supplied to the RV.

An additional (secondary) function of the vent valve is to prevent unacceptable bypass flow during normal operation. Each valve assembly includes two retaining rings with varying thicknesses that have integral threaded bosses at both ends to accept the jackscrews (see Figure 5-1 ). They are fabricated from AMS 5658 Type 15-5 PH stainless steel in the H1100 condition.

Controlled Document AREVA Inc. ANP-3479NP Revision 0 MRP-227-A ApplicanULicensee Action Item 7 Analysis for Three Mile Island Unit 1 Topical Report Page 5-2 CSS Vent Valve Bottom Retaining Ring Figure 5-1: A Schematic Representation of the Vent Valve Retaining Rings in B&W-Designed PWRs (1) (View from Inside the Core Support Shield)

The function of the retaining rings is to retain the vent valve body in the vent valve nozzle . The consequence of failure of a retaining ring or portion of a retaining ring is loss of support function for the valve body (

] Failure of a retain ing ring or portion of a retaining ring (

]

As of the evaluation date , there is no known cracking or failures of the vent valve retaining rings ; however there are several known instances of more susceptible types of martensitic PH stainless steel materials (e.g., Type 17-4 PH) in other components and systems failing .

Controlled Document AREVA Inc. ANP-3479NP Revision 0 MRP-227-A Applicant/Licensee Action Item 7 Analysis for Three Mile Island Unit 1 Top ica l Report Page 5-3 5.2 Evaluation Inputs This section will describe the quantitative inputs for the evaluation , including inputs such as flaw size , degraded material properties , and stresses .

5.2.1 Flaw Size Per MRP-227-A, the vent valve retaining rings are not screened as potentially susceptible to service-induced flaws (i.e., IASCC , SCC , or fatigue) . Therefore , the focus of this section is the potential for [

]

Review of available CMTRs for the vent valve retaining rings indicated [

] Information regarding the actual observed flaw sizes

[

]

5.2.2 Degraded Material Properties

[

1

Controlled Document AREVA Inc. ANP-3479NP Revision 0 MRP-227-A Appl icant/Licensee Action Item 7 Analys is for Three Mile Island Unit 1 Topical Report Page 5-4 The fracture toughness of the vent valve retaining ring material is expected to be

[ ] is a reasonable lower bound . In comparison , this fracture toughness is [

] after long-term irradiation exposure .

5.2.3 Stresses A stress analysis considering all portions of the retaining rings [

] (as discussed in Section 5.3.1) the available information indicates that the stresses during

[

]

5.3 Evaluation The results of the methodology utilized are organized into several conclusions as discussed in the following sections.

Controlled Document AREVA Inc. ANP-3479NP Revision 0 MRP-227-A Appl icanULicensee Action Item 7 Analysis for Three Mile Island Un it 1 Top ical Report Page 5-5 5.3.1 Likelihood of Failure As of the time this document was compiled , there is no known cracking or failures of vent valve retaining rings in the B&W-designed PWRs . This confirms that stresses sustained [

] given that the original retaining rings at TMl-1 are expected [

] To date ,

one of the vent valve assemblies has been replaced due to a locking device failure ; the remaining original vent valve retaining rings [

] No additional decrease in fracture toughness is expected during the period of extended operation for the originally installed vent valve retaining rings at TMl -1. Furthermore , the expected fracture toughness value for the vent valve retaining rings is [

] Due to the improbability of [

]

Based on the discussion above , failure of the vent valve retaining rings is not expected during the period of extended operation .

5.3.2 Effect of Failure on Functionality While failure of the vent valve retaining rings is not expected , this section describes the outcome, should a vent valve retaining ring fail , on the two functions of the vent valve retaining ring.

Controlled Document AREVA Inc. ANP-3479NP Revision 0 MRP-227-A Applicant/Licensee Action Item 7 Analysis for Three Mile Island Unit 1 Topical Report Page 5-6 One of the functions of the vent valves is to relieve pressure in the interior of the core support assembly during a cold leg large break LOCA. The retaining rings , if damaged due to TE (cracked , fractured material , surface irregularities , etc.), could eventually lead to [

] In this scenario , it is [

] during a large break cold leg LOCA. Therefore, degradation of the vent valve retaining ring material due to TE is not anticipated to affect the function of the vent valve during a cold leg large break LOCA.

An additional function of the vent valves is to [

] The event would be detectable

[

]

5.4 Conclusions The vent valve retaining rings at TMl-1 are not expected to fail during the period of extended operation based on the following:

. [

]

. [ ]

Controlled Document AREVA Inc. ANP-3479NP Revision 0 MRP-227-A ApplicanULicensee Action Item 7 Analysis for Three Mile Island Unit 1 Top ical Report Page 5-7

. [

]

. [ ]

In the unlikely event that failure of the vent valve retaining rings does occur, it is not expected to impair the function of the vent valve to relieve pressure in the interior of the core support assembly during a cold leg large break LOCA. Should the vent valve retaining rings fail and cause , [

] it would be detectable [

]

Therefore , the vent valve retaining rings at TMl-1 are expected to perform their function for the period of extended operation and in the unlikely event of failure , the primary vent valve function is not expected to be impaired and impairment of the secondary vent valve function would be detectable.

Controlled Document AREVA Inc. ANP-3479NP Revision 0 MRP-227-A ApplicanULicensee Action Item 7 Analysis for Three Mile Island Unit 1 Topical Report Page 6-1 6.0 OVERALL CONCLUSIONS A/LAI 7 is applicable to CRGT spacer castings, IMI guide tube spider castings , and vent valve retaining rings for TMl-1.

Based on the extensive evaluations summarized above , failure during the period of extended operation was found to be improbable for the CRGT spacer castings, vent valve retaining rings , and IMI guide tube spider castings. In the unlikely event of a failure occurring for these component items, the intended function of the component items is expected to be maintained or the failure will be detectable.

Controlled Document AREVA Inc. ANP-3479NP Revision 0 MRP-227-A Applicant/Licensee Action Item 7 Analys is for Th ree Mile Island Unit 1 Top ical Report Page 7-1

7.0 REFERENCES

1. Materials Reliability Program : Pressurized Water Reactor Internals Inspection and Evaluation Guidelines (MRP-227-A) . EPRI , Palo Alto , CA:

2011 . 1022863.

2. WCAP-17096-NP , Revision 2, "Reactor Internals Acceptance Criteria Methodology and Data Requirements ," December 2009 , NRC Accession No. ML101460157.
3. Letter AREVA-13-02949 , "AREVA Revised Text for Draft WCAP-17096-NP Revision 2 for Transmittal to EPRI to Address NRC Reviewer Comments ," October 23 , 2013 , NRC Accession No. ML16043A095.

4 . Letter from Christopher I. Grimes (NRC) to Douglas J. Walters (NEI) ,

"License Renewal Issue No. 98-0030 , "Thermal Aging Embrittlement of Cast Austenitic Stainless Steel Components ," May 19, 2000 , NRC Accession No. ML003717179 .

5. Kim , C., et. al. , "Embrittlement of Cast Austenitic Stainless Steel Reactor Internals Components ," Proc. 6th Intl. Symp . on Contribution of Materials Investigations to Improve the Safety and Performance of LWRs , Vol. 1, Fontevraud 6, French Nuclear Energy Society, SFEN , Fontevraud Royal Abbey , France , Sept. 18-22 , 2006.
6. NUREG/CR-7027, "Degradation of LWR Core Internal Materials Due to Neutron Irradiation," NRC ADAMS Accession No. ML102790482.
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